1. Field
This invention relates in general to monitoring the reactivity of a core of a nuclear reactor within the source range and more particularly to monitoring boron dilution during a reactor outage.
2. Related Art
In a pressurized water reactor power generating system, heat is generated within the core of a pressure vessel by a fission chain reaction occurring in a plurality of fuel rods supported within the core. The fuel rods are maintained in spaced relationship within fuel assemblies with the space between fuel rods forming coolant channels through which borated water flows. Hydrogen within the coolant water moderates the neutrons emitted from enriched uranium within the fuel rods to increase the number of nuclear reactions and thus increase the efficiency of the process. Control rod guide thimbles are interspersed within the fuel assemblies in place of fuel rod locations and serve to guide control rods which are operable to be inserted or withdrawn from the core. When inserted, the control rods absorb neutrons and thus reduce the number of nuclear reactions and the amount of heat generated within the core. Coolant flows through the assemblies out of the reactor to the tube side of steam generators where heat is transferred to water in the shell side of the steam generators at a lower pressure, which results in the generation of steam generally used to drive a turbine-generator set for the production of electricity. The coolant exiting the tube side of the steam generator is driven by a main coolant pump back to the reactor in a closed loop cycle to renew the process.
The power level of a nuclear reactor is generally divided into three ranges: the source or startup range, the intermediate range, and the power range. These three ranges can be further divided into six modes: mode one where the power is greater than five percent; mode two where the power is less than five percent; mode three, known as hot standby, where the temperature of the coolant is greater than 350°; mode four, also known as hot shutdown, where the temperature of the coolant is less than 350°; mode five, cold shutdown, where the temperature of the coolant is less than 200°; and mode six, refueling, where the temperature is less than 140°. The power level of the reactor is continuously monitored to assure safe operation. Such monitoring is typically conducted by means of neutron detectors placed outside and inside the reactor core for measuring the neutron flux of the reactor. Since the neutron flux in the reactor at any point is proportional to the fission rate, the neutron flux is also proportional to the power level.
Fission and ionization chambers have been used to measure flux in the source, intermediate and power range of a reactor. Typical fission and ionization chambers are capable of operating at all normal power levels, however, they are generally not sensitive enough to accurately detect low level neutron flux emitted in the source range. Thus, separate low level source range detectors are typically used to monitor neutron flux when the power level of the reactor is in the source range.
The fission reactions within the core occur when free neutrons at the proper energy level strike the atoms of the fissionable material contained within the fuel rods. The reactions result in the release of a large amount of heat energy which is extracted from the core in the reactor coolant and in the release of additional free neutrons which are available to produce more fission reactions. Some of these released neutrons escape the core or are absorbed by neutron absorbers, e.g., control rods, and therefore do not cause traditional fission reactions. By controlling the amount of neutron absorbing material present in the core, the rate of the fission process can be controlled. There are always random fission reactions occurring in the fissionable material, but when the core is shut down, the released neutrons are absorbed at such a high rate that a sustained series of reactions do not occur. By reducing the neutron absorbent material until the number of neutrons in a given generation equals the number of neutrons in the previous generation, the process becomes a self-sustaining chain reaction and the reactor is said to be “critical.” When the reactor is critical, the neutron flux is six or so orders of magnitude higher than when the reactor is shut down. In some reactors, in order to accelerate the increase in neutron flux in the shutdown core to achieve practical transition intervals, an artificial neutron source is implanted in the reactor core among the fuel rods containing the fissionable material. This artificial neutron source creates a localized increase in the neutron flux to aid in bringing the reactor up to power.
In the absence of an artificial neutron source, the ratio of the number of free neutrons in one generation to those in the previous generation is referred to as the “neutron multiplication factor” (Keff) and is used as a measure of the reactivity of the reactor. In other words, the measure of criticality for a nuclear core is Keff, that is, the ratio of neutron production to total neutron loss attributable to both destruction and loss. When Keff is greater than one, more neutrons are being produced than are being destroyed. Similarly, when Keff is less than one, more neutrons are being destroyed than are being produced. When Keff is less than one, the reactor is referred to as being “subcritical.”
During a refueling outage the reactor head is removed along with the upper internals structure to gain access to the fuel assemblies within the core. To provide such access, the control rods are removed with the upper internal structure, however, to maintain the fuel assemblies within the core subcritical with the control rods removed, the concentration of boron within the coolant within the core is increased. Monitoring the reactor coolant system boron concentrations during plant outages is a key aspect to verifying shutdown margin and preventing an inadvertent criticality. During the transition from mode six up to mode two, following a refueling outage, the plant will change boric acid concentration from approximately 2,300 ppm to 1,800 ppm. During this transition period in the outage, there are a large number of activities being performed which demand the attention of plant operators. Typically, shutdown margin is primarily monitored by periodic reactor coolant system boron concentration sampling and administratively locking out dilution sources to prevent inadvertent dilutions of boron in the reactor coolant system. Most plants also have boron dilution monitors in place that sample the output signals of the source range detectors for a statistically significant increase in count rates and provide an alarm based on a manually adjusted, preset increase in the count rate; e.g., a doubling in the count rate.
During a refueling outage at one commercial nuclear plant in May 2011, the reactor coolant system boron concentration unexpectedly decreased from 2,443 ppm to 1,483 ppm in 24 hours due to a leaking valve. The required shutdown boron concentration was 1,410 ppm. The source range detectors' output signals increased by a factor of three, but operations attributed this increase primarily to the concurrent reactor coolant system temperature increase. This utility had previously removed their secondary artificial neutron sources, which lower their source range detector count rate and make observing reactivity changes in a very subcritical state more difficult. While the plant was following the standard outage chemistry monitoring surveillance of taking samples every 24 hours, and the valves creating a dilution flow path were closed and locked out, but one leaked. A concurrent reactor coolant system temperature increase from about 325° F. to 475° F., while the dilution was occurring, masked the affect on the source range detectors and reduced the effectiveness of the manually adjusted, high flux at shutdown alarm.
Thus, a more sensitive detection system for identifying reactor coolant system boron concentration changes is desired.
Accordingly, it is an object of this invention to provide such a dynamic system that will continuously monitor boron concentration changes for changes that may not be readily detectable using the current methodology.
It is a further object of this invention to accurately monitor small changes in boron concentrations during operating modes six through two to protect a plant against an inadvertent criticality.
Further, is the object of this invention to provide such a system that is reliable enough to allow the elimination of the operating and core design constraints imposed by the current boron dilution accident analysis requirements.